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Title: CFD Analysis of Core Bypass Phenomena

Abstract

The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computationalmore » Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary« less

Authors:
; ;
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
974775
Report Number(s):
INL/EXT-09-16882
TRN: US1002586
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; COMPUTERIZED SIMULATION; COOLANTS; DESIGN; FLUID MECHANICS; FUEL PINS; GEOMETRY; GRAPHITE; INTERSTITIALS; IRRADIATION; LIFETIME; MANUFACTURING; PHYSICS; REACTOR CORES; SAFETY ANALYSIS; SYMMETRY; TEMPERATURE DISTRIBUTION; THERMAL EXPANSION; THERMAL HYDRAULICS; bypass flow; Computational fluid dynamics; NGNP

Citation Formats

Johnson, Richard W, Sato, Hiroyuki, and Schultz, Richard R. CFD Analysis of Core Bypass Phenomena. United States: N. p., 2009. Web. doi:10.2172/974775.
Johnson, Richard W, Sato, Hiroyuki, & Schultz, Richard R. CFD Analysis of Core Bypass Phenomena. United States. https://doi.org/10.2172/974775
Johnson, Richard W, Sato, Hiroyuki, and Schultz, Richard R. 2009. "CFD Analysis of Core Bypass Phenomena". United States. https://doi.org/10.2172/974775. https://www.osti.gov/servlets/purl/974775.
@article{osti_974775,
title = {CFD Analysis of Core Bypass Phenomena},
author = {Johnson, Richard W and Sato, Hiroyuki and Schultz, Richard R},
abstractNote = {The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary},
doi = {10.2172/974775},
url = {https://www.osti.gov/biblio/974775}, journal = {},
number = ,
volume = ,
place = {United States},
year = {2009},
month = {11}
}